ES2758707T3 - Installation and procedure for waste treatment - Google Patents

Abstract

Facility for the treatment of activated, contaminated and / or uncontaminated waste, in which, in a modular type of construction with low radiation exposure and low transport costs, activated, contaminated and / or uncontaminated, decoupled waste or waste of the decommissioning and regardless of a particular nuclear installation, they are supplied to a recycling station, a landfill or a waste deposit after separation and / or decontamination, consisting of - a contact surface in a building (27) with a entrance and an exit (27.2, 27.3) for personnel, - at least one access for material, which is made as a sluice (1, 1.1, 1.2) for contaminated parts, - at least one exit (27.4) for material (31) to be delivered, which is decontaminated, - at least one outlet (1, 1.1, 1.2) for radioactive material or waste that can be released in a restricted way, - treatment stations (12, 13, 14, 15, 16, 17, 18, 19, 20, 21, 22), which they are separated from each other by flexible and adjustable walls (28), so that their surface can be increased, reduced or modified with little effort according to the required performance, - treatment stations (2, 3, 4, 5), which are separated from each other by solid walls (25), as well as - transport containers, means of transport and spaces and storage rooms, the treatment stations, in which a high dose rate occurs, being arranged in a high-pressure area dose rate far removed from the entry and exit of personnel (27.2, 27.3) and the exit (27.4) for decontaminated material (31) to be delivered.

Description

DESCRIPTION

Installation and procedure for waste treatment

The invention relates to an installation for the treatment of activated, contaminated and / or uncontaminated waste corresponding to the characteristics of the first claim and to a method for the treatment of the mentioned waste.

The facility and procedure can be used anywhere where metal and mineral wastes as well as combustible and non-combustible wastes must be machined from dismantling a nuclear facility so that, with a flexible modular type of construction of individual treatment stations and With low radiation exposure and low transport costs, contaminated waste should be supplied, after separation and decontamination, on the one hand to recycling and, on the other hand, to a landfill or nuclear cemetery.

In the case of the decommissioning of a nuclear power plant, metallic and mineral waste are produced, as well as combustible and non-combustible waste. Of the total existing mass of a nuclear power plant (without fuels), approximately 10% must be treated, the rest can be supplied for free use. According to the power plant, the parts of the installation, waste and secondary waste to be treated can be between 200001 and 50,000 t. Most of these materials, about 70%, are fed either through direct decision metering or through melting for unrestricted release. The remainder, approximately 30%, may be disposed of in part in landfills (building materials and structural substances), but in another part it must be supplied to a nuclear cemetery as medium to low-level radioactive waste.

Additionally, large components such as, for example, steam generators can be processed. To achieve the desired waste disposal goals, some components must be treated multiple times, based on a return rate of approximately 20%. In general, only materials whose waste disposal route has been clarified can be delivered.

Different facilities and procedures are known for the treatment of contaminated waste.

As long as the surfaces of contaminated components are radioactively contaminated, EP 0 638 516 A1 describes a procedure for removing lead and cadmium from phosphoric acid. This procedure is useful when performing chemical or electrochemical decontamination and spent phosphoric acid must be regenerated from purification of contaminated parts.

WO 2013/175089 A1 describes a chemical procedure for decontaminating steel parts with phosphoric acid solution.

US 5,037,159 describes the removal of a contaminated layer by knocking off contaminated concrete pieces.

DE 2638174 A1 describes the decommissioning of a nuclear reactor.

By document XP 055402261, D. Rittscher et al :. "Kernkraftwerke Greifswald und Rheinsberg: Stilllegung, Abbau und Aufbau", International Review for Nuclear Power, Vol. 47, No. 4 (2002-04-01), pages 235 to 243, there is also known a facility for the treatment of activated, contaminated and / or uncontaminated waste, namely the Zwischenlager Nord (ZLN). However, the ZLN is directly assigned to a nuclear facility, namely the former Greifswald nuclear power plant.

Regardless of this, different waste treatment procedures are known, which are used to transport, classify and / or crush the waste.

In the known procedure, it is disadvantageous that the waste treatment is not carried out in a decoupled manner from the dismantling of nuclear facilities, but directly on site. With this, a facility for the treatment of contaminated waste can only be used for the dismantling of a single facility and, correspondingly, its capacity and space conditions.

In these cases, planning and coordination activities must be managed for each individual installation. Due to the existing premises, there are restrictions regarding the treatment surface. As a general rule, optimal transport routes to treatment stations are not available. A modular structure is neither necessary nor possible, since the planned and completed installation is planned only for the specific power plant to be dismantled. As a general rule, synergies regarding interchangeability of personnel between locations are not possible.

Therefore, the objective of the invention is to develop an installation and a procedure for the treatment of contaminated waste, which is permanently installed in a location intended for it and, therefore, is decoupled. dismantling, individual treatment stations can be reduced or increased with a modular principle, no mutual interference of transport routes being present, with high-quality groups and personnel familiar with the production of a minimum of radioactive waste, so that a high percentage of waste can be supplied to recycling.

This objective is achieved by means of an installation according to the characteristics of the first claim and a method according to the characteristics of claim 8. The secondary claims express advantageous configurations of the invention.

The solution according to the invention provides for a facility for the treatment of contaminated waste, in which this waste is mechanized in such a way that, as far as possible, a high percentage of the waste is supplied for recycling.

The facility consists of a contact surface with a building with an entrance and exit for the personnel, a lock for contaminated parts and an exit for decontaminated material, with treatment stations arranged on the contact surface, which are separated from each other by flexible walls. On the contact surface are transport containers, means of transport and storage rooms. The treatment stations, in which a high dose rate occurs, are arranged in an interval for high dose rates, which is arranged far from the entrance and exit of the personnel as well as the exit for releasable material.

In addition, a cleaning room, a service room, an office and measurement rooms, as well as an oil and liquid storage room are arranged on the contact surface.

It is advantageous if the walls of the building are provided with an easily decontaminable covering. In the building is the contact surface on which solid and flexible walls are arranged. Solid walls represent walls of bricks, concrete, or other components. Solid walls are arranged around treatment areas or stations whose surface is unchanged. Flexible walls can represent plastic sheeting, tarps, tents, partitions, or other adjustable walls.

Due to the flexible walls, the treatment stations can be expanded, reduced or, if necessary, modified accordingly to the required performance.

The treatment stations comply with the different processing requirements regarding the required area and the procedure to be carried out for the contaminated waste.

In this regard it can be a disassembly area for large components, a concrete or concrete surface treatment, a shot blasting preparation, a cable treatment or processing, a thermal fragmentation, a waste sorting radioactive and non-radioactive, a filling and measuring facility, a sorting facility, a washing facility, a drying facility, a high-pressure press, a fusion container optimization, a manual fragmentation, a wet chemical decontamination, a facility wheel spinner, shot blasting facility, ultrasonic bath, box washing facility, sheet metal processing, asbestos conditioning or hazardous facility, container filling or evaporation facility.

The used transport containers can represent barrels, small or large containers, transport cages, POLLUX, in particular Konrad containers, or containers.

As means of transport, electronic stacker trucks, forklifts, gantry cranes, jib cranes, transport equipment or suspended monorails are used, which are used selectively in the various areas.

In the case of decommissioning of nuclear power plants, a large amount of contaminated material is produced, mainly metal and concrete. To reduce the volume of radioactive waste, disassembled contaminated material should be divided or separated into released material, which is treated with decontamination methods, and radioactive waste for long-term storage. State inspection bodies establish the limit value up to which a liberatory measurement of contaminated material is allowed. For example, the limit value up to which a liberatory measurement of the material from a decommissioned reactor is possible may be 0.1 becquerel / gram (Bq / g).

Usually there is a contamination with uranium oxide, plutonium oxide, strontium-90, cesium-137 and cobalt-60. The radioisotope content of the waste is determined by radiochemical nuclide vectors that are extracted from samples of the contaminated material. Nuclide vectors are determined by the proportion of radioactive radionuclides to be measured, such as, for example, cobalt-60 and cesium-137. It is then assumed that the radioisotopes respectively have the same dispersion within the individual areas of each decommissioned nuclear power plant. For example, an abandoned nuclear power plant can be divided into 10 areas with their own nuclide vectors respectively. Thereafter, a radioisotope is measured and the remaining radioisotopes of the waste are assumed to be distributed as determined in the nuclide vector of the corresponding area of the nuclear power plant. In the current representation, this is cobalt-60. This means that large components, metal debris, concrete slabs, concrete debris and other debris must be removed in particular.

It is advantageous to have at least one band saw, at least one hydraulic shear or at least one arc saw in the cutting area for large components.

Furthermore, it is advantageous to use in concrete processing at least one core drill bit, at least one concrete milling machine, at least one circular saw for large or small concrete slabs, arranging at least one wire saw or at least one concrete crusher .

Thermal fragmentation takes place in submersible drawers. A submersible drawer is an installed movable wall enclosure with a roof cover. Access for personnel is through locks. The transport of material is carried out through openings of doors, gates or roofs. In this case, a lock function does not necessarily have to be fulfilled. The transport of material is carried out with installations of cranes, forklift trucks or forklift trucks. In these, a welding table can be arranged as well as a parking space for transport cages. In the installation, three submersible drawers are advantageously arranged for thermal fragmentation.

In the dry blasting installation, a cavity blasting installation, a blasting box and / or a spinning wheel installation and a shot blasting preparation for jet cleaning may be provided. Cables can be machined or processed in cable processing or preparation by cable crushers, plastic crushers, and cable strippers.

Furthermore, a box washing installation may be arranged.

In manual fragmentation, thin plate processing may be arranged.

Furthermore, it is advantageous to provide in the waste sorting a container for radioactive material and fuel, a sorting for non-radioactive combustible material and its storage, a sorting of waste for radioactive waste and a binding machine as well as a packing press and a press inside of the barrel. Furthermore, it is advantageous to provide, in a drying installation, a drying oven, a cobalt (CC) coincidence measurement installation, a control for the control apparatus and a barrel measurement.

Furthermore, it is advantageous to provide, at the keg bottling facility, a keg supply, a keg bottling, a keg dispensing, a high pressure press, a hydraulic and control container for controlling the keg bottling facility and a beading machine.

The measuring room should be provided with instruments to measure and record the dose rates such as for the identification of the measured containers. Furthermore, it is advantageous to provide wet decontamination of hot and cold water under high pressure as well as to provide in this area a bias cut belt filter, a pump filter for a first intermediate container and a pump filter for a second intermediate container.

In the treatment of asbestos or the conditioning of asbestos, a press can be arranged inside the barrel.

The facility for the treatment of contaminated waste has the advantage that the treatment of the waste is carried out in a decoupled manner from the dismantling and independently of a specific power plant. In this installation, mutual interference through the transport routes is not possible.

The modular principle makes it possible to easily adapt the equipment to the dismantling needs, and the installation can be adapted to the treatment capacity required during dismantling.

Since the installation is designed in such a way that the treatment stations are arranged on the contact surface in a manner corresponding to the dose rate, the burden on the environment and people remains low.

The facility allows recycling to be provided to the highest possible percentage of contaminated waste.

The installation also makes it possible to avoid unnecessary radioactive transport, particularly in the national territory. Through special on-site procedures, highly effective decontamination and optimal fragmentation can be performed, whereby maximum utilization of the packaging barrels of the nuclear cemetery occurs by combining types of waste.

The procedure according to the invention for the treatment of activated, contaminated and / or uncontaminated waste from the dismantling of nuclear facilities envisages supplying the majority of the waste for recycling, these being mechanized in one or more treatment stations, after of which a decisive measurement takes place and the decontaminated waste is supplied to a recycling facility, the contaminated waste to a landfill and the radioactive waste to a nuclear cemetery.

The treatment of metallic waste, which can also represent large components such as steam generators, first foresees manual, thermal and / or mechanical fragmentation, then decontamination by high pressure cleaning or chemical surface treatment or a dry blasting, after which a pre-measurement or decision measurement takes place. Wet chemical surface treatment can take place by a procedure such as that described in WO 2012/175098 A1. After pre-measurement and decision-making measurement, it is evident whether decontamination by procedures The aforementioned was successful and the metallic waste can be supplied for recycling or if the decontamination stages must be repeated. After the decisive measurement, the metallic waste can leave the installation or, in exceptional cases, must be treated repeatedly.

Mineral wastes such as concrete pieces and structures are removed first. This disassembly can be done with a wire or circular saw or with pneumatic or hydraulic hammers or other suitable devices. Parts such as, for example, displacement parts or other steel parts sealed in the concrete are removed, which prevent decontamination. The contaminated surface of concrete plates is milled. The concrete plates and pieces treated in this way are supplied to a pre-measurement and decisive measurement, with an assignment being made to conventional or radioactive waste, and then these pieces leave the installation by the intended route.

Debris such as concrete pieces are machined by treating the concrete surface using a concrete milling machine that strips the contaminated debris to a required depth. The resulting dust particles are sucked in by negative pressure and supplied to a dust separator, for example a cyclone. The negative pressure is generated by a vacuum pump above the cyclone. At least one filter is arranged between the cyclone or the dust separator to separate fine dust particles. Furthermore, it is advantageous to arrange a safety filter after a first filter. The filter after the cyclone is cleaned by emptying the filter cake down through the cyclone. The dust collected in the filter is monitored by means of a detector for its radioactive components and cleaned if the corresponding limit value is exceeded.

In the dust separator the powdery particles are separated, which leave the separator downwards, where they are collected in a suitable container. From the dust particles leaving the separator, the radioactivity is continuously measured by an activity measuring device. Depending on the radiation exposure to the dust, it is either removed as dust giving liberatory measurement or supplied to a nuclear cemetery as contaminated dust. Uncontaminated dust can be used. The separation of the contaminated and uncontaminated dust is carried out after the activity measurement by means of a two-way valve. The transport of dust particles can be carried out by means of suitable containers such as barrels.

The device for treating concrete pieces and the procedure for treating these concrete pieces are described in detail in an exemplary embodiment. The devices and procedures can also be leveraged regardless of the current entire facility as well, separately.

The treatment of mineral waste such as rubble is carried out in such a way that it is first pre-measured, after which a division into conventional or radioactive waste is carried out. Radioactive and also conventional wastes are separately screened, the coarse fraction respectively being crushed from a subsequent coarse crushing in a crusher such as a jaw crusher to a grain size intended for it. In this regard, conventional waste in other sorting and shredding stations is treated as radioactive waste, ie conventional waste and radioactive waste are respectively supplied separately for sorting and shredding. The shredded waste is again supplied to a decision-making measurement, after which the conventional waste is packaged and supplied to a landfill, and the radioactive waste is either filled in a known way in a Konrad container or pressed and then left in the facility. towards a nuclear cemetery.

Debris such as cables are first subjected to decontamination by scrubbing or other suitable procedure, after which a stripping of the cable and, in another process step, the crumbling of the cable is performed. Also, the stripped plastic parts of the cable may crumble. It is advantageous to separate the cables and the separate plastic parts separately. The peeled and shredded parts of the cable are supplied after crushing to a decisive measurement, after which a packaging and recycling of the reusable waste is carried out or, whenever there are contaminated parts, a filling in a waste container with the subsequent transport to storage in a known manner.

The treatment of the waste is carried out in such a way that it is first classified in the installation into conventional and radioactive waste, which can be done on the basis of a decisive measurement. Conventional combustible wastes are pressed and supplied to a preliminary and decisive measurement, after which the uncharged wastes are supplied to external conventional combustion. Radioactive waste is separated into combustible and non-combustible waste, the combustible radioactive waste being pressed and supplied to a dose rate measurement and contamination measurement with subsequent external combustion. Non-combustible radioactive waste is supplied to a high pressure press and to drying.

Burned and pressed and dried radioactive waste is supplied to a storage, optionally barrels or containers, and to an evacuation.

Next, the installation will be explained in more detail in ten figures and an example of embodiment. The figures show:

Figure 1: installation for the treatment of contaminated waste in a schematic representation, the treatment stations being represented on the contact surface.

Figure 2: facility for the treatment of contaminated waste with the procedure route from a container that delivers metal parts, which are decontaminated at the facility. Figure 3: Installation for the treatment of contaminated waste in a schematic representation and with the route of the waste that is supplied to a classification in the installation.

Figure 4: Installation for the treatment of contaminated waste in a schematic representation and the route of large components, which are disassembled and decontaminated in the installation.

Figure 5: Installation for the treatment of contaminated waste in a schematic representation and the route of concrete pieces through the installation, which are machined in the installation. Figure 6: section through the side view of a nuclear power plant

Figure 7: Partial section through a concrete plate such as a concrete pavement

Figure 8: Representation of procedure steps 65-71

Figure 9: Representation of a device for separating radioactively contaminated material and its treatment Figure 10: Detailed representation of a cobalt coincidence measuring apparatus

Figure 1 shows the installation according to the invention for the treatment of contaminated waste in a schematic representation, consisting of building 27, in which the contact surface is located, which has locks 1.1, 1.2, through which Contaminated parts such as a steam generator 32 or small contaminated parts arrive at building 27 individually or in a 20 'container. For staff, a 27.2 entrance and exit are arranged for people in the upper right and 27.3 emergency exits. Decontaminated material can again leave building 27 through the decontaminated material exit 27.4 in the upper right. Contaminated material can leave building 27 again through locks 1.1, 1.2. Different treatment stations are arranged on the contact surface 29, which are called. Flexible walls 28 are arranged around the treatment station. These flexible walls may represent, for example, a tent or other suitable device. In this way, it is possible to increase, decrease or modify the treatment station with a little effort within the contact surface. In addition, on the contact surface there are means of transport, an office and a cleaning and service room 23.1, an oil and liquid warehouse 23.2, a locker for the distribution of tools 23.4, stacker and loading truck surfaces 23.3. Transport containers such as a barrel, transport cages, POLLUX or containers are placed on the contact surface.

As treatment stations on the contact surface 29 can be the cutting area for large components (treatment of large components) 15, a treatment of the concrete surface 22, a band saw 16, a thermal fragmentation 12, a grinder concrete 21, a manual fragmentation 13 with fine sheet metal processing 14, a mechanical-thermal fragmentation, a load optimization, a DC measurement 25.2, a wet contamination with chemical agents 10 or a high pressure water jet 8, a treatment of cables 17, a sorting tent 20 for non-radioactive waste, a sorting tent 19 for radioactive waste, an asbestos conditioning 18, a barrel drying / measuring facility 3, a high pressure press 2, an ultrasonic bath 9 , a Konrad 4 container load.

Corresponding to the dose rates present in the different areas, the contact surface on the right side presents an area with treatment stations with a low dose rate, in the center an area with treatment stations with a medium dose rate and on the left an area with treatment stations with a high dose rate. This has the advantage that the heavily contaminated parts are in one area and the least contaminated parts are in the other area. The high dose rate area is in front of the entrance and exit 27.2 for people and exit 27.4 for decontaminated material. Parts with a high dose rate are in the high dose rate area and can also leave the contact surface in this area without reaching the other areas.

Figure 2 shows the route of contaminated and decontaminated metal or steel parts 30, 31 through the installation according to the invention. Delivery is made through the inner and outer sluice 1.1, 1.2 in a 20 'container. The metal parts 30, 31 are removed from it by a suitable transport apparatus and supplied for manual fragmentation 13, sheet metal processing 14 or thermal fragmentation 12. In this working area, the metal parts 30, 31 can unscrew and provide transport rings. In the subsequent mechanical-thermal fragmentation, the pieces can be disassembled, for example, by sawing or separating cutting. This fragmentation can be done automatically. The metal parts are then transported through the facility for decontamination. Depending on the size of the part, this can be done using electronic stacker trucks, forklift trucks, or gantry cranes. Decontamination can be performed in a known manner by high pressure cleaning, a dry blasting procedure, a cavity blasting procedure, a spinning wheel procedure, or a shot blasting box procedure. Decontamination by means of phosphoric acid with an action time of 5 to 6 hours is carried out in the installation and in Figure 2 in a wet decontamination 10.

After wet decontamination 10, the treated parts are stored on a buffer surface 24 and supplied to a pre-measurement area 26 for pre-measurement and final measurement, after which the impeccably decontaminated metal steel parts 31 leave the facility. Steel parts After the contamination measurement 25.2, the contaminated 30 are transported through the locks 1.1, 1.2, if appropriate, by means of containers outside the building 27 or are again decontaminated.

Figure 3 shows the waste treatment facility and the route of waste through this facility. The delivery of waste such as contaminated waste 33 is carried out through the locks 1.1, 1.2 at the facility, in this case by a 20 'container. This is transported to the radioactive waste sorting tent 19 and is classified into conventional and radioactive waste. Uncharged waste is supplied to conventional combustion. The radioactive waste is separated into combustible and non-combustible waste, the combustible radioactive waste being pressed and then supplied to a combustion and non-combustible radioactive waste to a high pressure press 2 and drying 3, after which the radioactive waste is they fill in containers (Konrad) 4 of nuclear cemetery and are supplied through locks 1.1, 1.2 to a nuclear cemetery.

Figure 4 shows the waste treatment facility and the path of a steam generator 32 through this facility. The steam generator 32 and its parts are first disassembled in the fragmentation of large components 15. For this purpose, band saws 16 are arranged, which break up the steam generator 32 into easily transportable parts. Another fragmentation is performed as a thermal fragmentation 12, in which the steam generator 32 or individual parts are broken into smaller parts by means of separation welding. The individual parts of the steam generator 32 are then supplied at a dose rate measurement 25.2.

The contaminated steel pieces 30 leave the facility through the locks 1.1, 1.2. The decontaminated parts are supplied to the wet decontamination facility 10, where decontamination is performed. The machined parts are temporarily stored on a buffer surface 24. Parts that are decontaminated in a manner corresponding to the decisive measurement in the premeasuring area 26 can leave the facility via the outlet for decontaminated material 27.4 towards recycling.

Figure 5 shows the waste treatment facility and the mineral waste route 37 as concrete pieces through this facility. Mineral wastes such as concrete pieces, for example concrete slabs, representing building elements in a power station building, are delivered by means of containers through the inner and outer sluice 1.1, 1.2, discharged and they are transported by means of a gantry crane to the concrete surface treatment 22. The contaminated surfaces can be worn away in the treatment of the concrete surface 22, the contaminated worn surfaces being supplied to the container load (Konrad) 4, while that the decontaminated concrete slabs or plates 38 leave the installation again due to the treatment of the concrete surface 22 and the outlet for decontaminated material 27.4. Other small pieces of concrete can be crushed by concrete crushers 21, for example, jaw crushers, and, whenever they are contaminated, they are filled in the Konrad 4 container load into nuclear cemetery containers and leave the facility through the locks 1.1, 1.2 in the direction of the nuclear cemetery. Broken and decontaminated concrete pieces 39 are supplied for recycling and leave the facility through the decontaminated parts outlet 27.4.

More possible processing routes through the facility are conceivable, for example for the treatment of cables and other parts, which occur during the dismantling of a power plant and should be exploited.

Figure 6 shows the cross section of a nuclear power plant 50 that should be dismantled. Different nuclide vectors may be associated with the individual groups of metal parts within the safety container 51. Another nuclide vector may be associated with the z-shaped concrete deck slabs 52, which form the cover of the safety container 51. Another nuclide vector may be associated with pump chamber 53, in which there are machines and containers with the help of which water is pumped through the secondary cooling system. A frequent cause of radioactive contamination is a leak in the secondary refrigeration system, for example, by repairing or replacing pumps 54. Contaminated water runs through the floor 55 of pumping chamber 53 and flows below the pumps, toward corners and into cracks and cavities in the concrete floor. Figure 6 shows a conceptual cross section of a decommissioned nuclear power plant 50.

To facilitate cleaning and decontamination, the floors 55, walls, and z-shaped concrete deck slabs 52 of nuclear power plant 50 are typically provided with a two-component epoxy coating or urethane coating. The walls are usually covered up to a few meters in height. After a leak, for example, the floor may dry out. In concrete, however, cracks can form and depressions can occur due to falling parts. Contamination cannot always be removed from interstices and cracks.

Figure 7 shows a cross section of the concrete pavement 56 in which a crack has formed after the floor was originally provided with a decontamination varnish 58. After the formation of crack 57 in the ground, the liquid Radioactive leaked and reached crack 57. Although the contaminated site was almost completely cleaned up, contamination at the deepest site in crack 57 could not be removed. As usual in such cases, the concrete floor 56 was provided with a new coating 60 to seal contamination and ensure future cleanliness. During the life of the nuclear power plant 50, various coatings were applied to the concrete floor 56, including cover layer 61. The z-shaped concrete walls and cover slabs 52 of nuclear power plant 50 are similarly coated.

When a nuclear power plant is decommissioned, contamination in the concrete is usually only present up to a surface depth of approximately 4 mm. By peeling off the outer layers to a thickness of about 10 mm, the contaminated percentage can be fundamentally completely removed. The outer layer of the concrete floor, walls and beams are peeled off by sanding or stripping. In one type of knock-off, the concrete surface is subjected to a treatment with thousands of hammer blows per minute, and therefore the deck surface is sprayed; this dust is then sucked up. A more detailed description of removing the contaminated layer by knocking off is mentioned in US Patent No. 5,037,159 to Nutter et al., Which is referred to herein in this case.

Although only about 1 cm is released from the concrete and coating, a large amount of possibly contaminated waste and contaminated dust is produced in this regard. In one embodiment, approximately 200 kg of powder is produced per hour. By removing the top layer of concrete and determining how much of the resulting concrete and paint particles in this regard exceed the limit value allowed for the release measurement, the amount of radioactive waste to be permanently stored can be significantly reduced. Since in an exemplary embodiment according to FIG. 7, for example, the top 10 mm of the concrete is shed from right to left, it is advantageous to separate the uncontaminated concrete powder from section 62 from the radioactively contaminated powder from section 63. However, there is currently no known procedure by which it can be determined whether the absorbed dust particles exceed the limit value for radioisotopes to separate them during blowing and to classify them accordingly in containers for releasing or waste measuring material. radioactive. The dust is usually stored in principle as radioactive waste. Known procedures for measuring radioactivity are not fast enough to measure limit values on dust particles as they fly through the aspirator tubes. Furthermore, the detached particles that are transported through the vacuum cleaner do not have enough mass to be measurable with conventional procedures. The present invention offers the possibility to measure the radioactivity of the powder and subdivide it accordingly in the barrels for material that gives liberatory measurement and radioactive waste.

Figure 8 shows in a flow chart of steps 65-71 of a procedure in which the powder that exceeds the limit value is separated from the powder that is below the limit value. In this embodiment, the limit value up to which the material is released for routine waste disposal is 0.1 Bq / g (for Co-60).

In a first stage of flow diagram 65, the concrete, which is contaminated with radioactive particles, is processed by sanding or chiselling to form powder. This dust occurs, for example, when coating 60-61 is removed from the concrete surface of concrete floor 56 or from the z-shaped concrete deck slab 52 when a nuclear power plant 50 is dismantled. The dust contains both concrete and paint particles.

In step 66 of the flowchart, a negative pressure is created, which draws the powder first into a hose 87 and then into a dust separator 81. The negative pressure is generated, for example, by a vacuum pump 72, which it is spatially arranged at the other end of the vacuum line. Dust is sucked during sanding of the coating.

Figure 9 shows a system for the separation of radioactively contaminated material 73 and illustrates the machines that are used to carry out the procedure described in figure 8. In figure 9 a manual concrete milling machine 74 is shown, which detaches the layer Concrete ground cover with the help of three diamond-ground conical molded cup wheels. In one embodiment with a manual 74 concrete milling machine, a PG820 from Husqvarna is used. In another embodiment, a chassis concrete milling machine, called The Moose, 76, manufactured by the Pentek company is used. The chassis concrete milling machine 76 removes the covering layer of a concrete floor with the help of oscillating chisels equipped with hard metal 77. The floor can be sanded or chiselled in strips 30 cm wide. The resulting powder in this regard is sucked into a vacuum hose 78, the suction opening 79 of which is positioned close to the sanding or chiselling job site. The powder reaches through the vacuum hose 78 towards the inlet of the dust separator 80 of the dust separator 81. In the embodiment of Figure 9, the dust separator 81 is a cyclone separator or filter. Figure 9 shows a sectional drawing of a dust separator 81 with a cylindrical overlay 82 protruding from the upper end of the separator. The aspirated air flows through the inlet of the dust separator 80 towards the dust separator 81 and exits through the cylindrical insert 82. The dust separator 81 has a cylindrical lining 84, a conical bottom part 85 and does not have Mobile parts. Due to the centrifugal force produced by the tangential velocity of the air during the downward movement along the interior walls of the dust separator 81, the dust particles 87 are separated from the air stream 83. In some types of embodiments , a worm-shaped baffle is placed around the outer side of the cylindrical overlay 82 at the upper end of the cylindrical liner 84 to direct the flow of the incoming air stream 83 toward a downward spiral. Powder 87 is passed through the bottom opening in the conical bottom, while air stream 83 rises through the center of powder separator 81 and leaves it through cylindrical overlay 82. The Dust separator 81 has a lower conical collection container 86 with a narrower bottom opening, through which dust 87 falls into one of the two collection containers 88/89 little ones.

Dust separator 81 is well suited for filtering contaminated concrete dust from air stream 83, since dust separator 81 does not have a filter membrane that must be emptied or clogged. Since the dust separator 81 has no moving parts, it can be operated for a long period without maintenance work. Although the dust separator 81 filters most of the dust 87 from the air stream 83, many dust particles less than 0.5 mm in diameter travel upward with the air stream 83 and through the cylindrical overlay 82 A filter, such as a ceramic recoil spark plug filter 90, is arranged on the cylindrical insert 82 and is used to filter out dust particles with a diameter of less than 0.5 mm. Air stream 83 flows through the bottom of the ceramic recoil spark plug filter 90 and exits from the sides of the cylinder. Therefore, small particles of the powder 87 remain on the inner side of the cylinder. A fine particle cake is formed on the filter element. As the cake thickens, the lumps of powder 91 break and fall back through the powder separator 81 and into the small collection containers 88/89, which are filled with the larger powder particles.

During operation, the recoil spark plug filter 90 is cleaned by an opposite blow of air pressure. This process removes all the fine particle cake from the filter element. The cake, consisting of lumps of powder 91, falls to the bottom of powder separator 81 and into small collection containers 88/89. Filter cleaning is carried out periodically to keep the ceramic recoil spark plug filter 90 clean.

In step 67 of the flow chart (Figure 8), a small collection container 88 or 89 is filled with the powder 87 and 91 that falls from the powder separator 81.

During filling, both small collection containers 88 and 89 are positioned under the dust separator 81 with the help of a conveyor kit 75. Each small collection container 88 and 89 has a volumetric capacity of approximately 10 liters, which corresponds to approximately 7 kg of concrete powder. This in turn has a density of approximately 700 kg / m3. It takes just over 2 minutes to fill the two 10-liter containers 88 and 89 when using a chassis 76 concrete mill or a manual 74 concrete mill, which generates approximately 200 kg of concrete powder per hour. As soon as a small collecting container 88 or 89 is filled with concrete powder, it is, for example, transported with the aid of conveyor equipment 75 to an activity measuring apparatus 92; the other small collecting container 89 is positioned below the dust separator 81, so that it can begin to be filled. In Fig. 9, the small collecting container 89 is filled with dust and has been brought to the activity measuring apparatus 92. The small collecting container 88 is filled with dust 87 and 91.

In step 68 of the flowchart (FIG. 8), the radioactivity of the powder in the small collecting container 89 is measured with the aid of the activity measuring apparatus 92. The measuring apparatus 92 takes advantage of the matching procedure to determine the activity of the cobalt radionuclide -60 in 87 and 91 powder.

FIG. 10 depicts in more detail the cobalt match measuring apparatus 92 which operates according to the matching procedure. Activity measuring apparatus 92 has two sets of oppositely arranged gamma detectors, which are aligned 90 degrees offset from each other. A gamma ray detector consists of a plastic scintillation detector 93 combined with a rapid photomultiplier tube (PMT) 94, which amplifies the light emitted by the plastic scintillation detector 93. The plastic scintillation detector 93 is equipped with a glass that emits light as soon as gamma rays arrive. The light intensity is proportional to the strength of the gamma radiation. Plastic scintillation detector 93 is coupled to fast photomultiplier 94, which converts light to electrons and then amplifies the electrical pulses generated by the electrons. A second gamma ray detector is located directly in front of the first gamma ray detector and is equipped with another plastic scintillation detector 95 and another rapid photomultiplier 96.

The second set of gamma detectors is aligned offset 90 degrees from the axis of the first set of gamma detectors to achieve a wider coverage of the 360 degree emission range. In one embodiment, the activity measuring apparatus 92 is a modified version of the large capacity monitor (LAM12) from the manufacturer Thermo Scientific of Erlangen. Activity measuring apparatus 92 measures activity by counting and determining the angular correlations of two consecutive gamma rays that are emitted by each isotope of cobalt-60 present in the powder. The matching procedure records gamma radiation (y1) from the fragmentation of a cobalt-60 isotope in the plastic scintillation detector 93 and other gamma radiation (y2) from the same fragmentation in the plastic scintillation detector 95. The measuring apparatus Activity 92 then only counts the matches that were also recorded simultaneously by the two opposing plastic scintillation detectors 93, 95. In this way, the background radiation can be recognized and calculated accordingly in the evaluation. The number of cobalt-60 isotopes per second is recorded and named in becquerel (number per second).

However, not all pairs of gamma rays from the fragmentation of cobalt-60 isotopes are detected as knocks by the activity measuring apparatus 92. The probability that a pair of emitted gamma rays is detected as knocks is equivalent to the probability of response of activity measuring apparatus 92 or detector; this is indicated as a percentage. When determining whether the limit value has been reached, the detector response probability must be taken into account. For example, if the response probability of the activity measuring device (detector) is 40%, then the number of strokes must be multiplied by 2.5 to calculate the actual number of strokes per second.

The cobalt matching procedure for measuring radioactivity is faster than other procedures and can determine during filling of the small collecting container 88 whether the powder in the first filled small collecting container 89 can be released for general disposal, since the powder has nuclear fragmentation below the release limit of 0.1 Bq / g or 100 Bq / kg. Each of the small collection containers 88 and 89 has a volumetric capacity of approximately 7 kg of powder, so that the release limit is exceeded when the activity measuring device 92 determines an equivalent adjusted by the probability of response of more than 700 actual strokes per second or 7000 strokes per 10 seconds. Since it takes 2 minutes to fill each 10-liter container 88 and 89, the activity measurement can be carried out within the interval of 10 seconds, within the period before the small collecting container 89 must return to the position below of the dust separator 81 to house the next quantity of powder 87 and 91.

In step 69 of the flowchart (FIG. 8), powder 87 and 91 is transferred from the small collection container 89 to a waste container 97 when the activity detected within a predetermined period of time exceeds a predetermined limit value. In a first embodiment, the time period is set to 10 seconds; the predetermined limit value is 0.1 Bq / g or 7000 strokes within these 10 seconds with 7 kg of powder in the small collecting container 89. In a second embodiment, the predetermined period of time is 1 second; the default limit value is 700 strokes within every second with 7 kg of powder in the small collecting container 89. In the second embodiment, the activity measurement is carried out ten times and the predetermined limit value is considered to be exceeded if the actual strokes are greater than 700 during the average of ten predetermined time periods.

In one embodiment, the waste container 97 is a 200 liter volumetric metal barrel, which can accommodate the volume of 20 small collection containers 88 and 89 with radioactive powder. FIG. 9 shows the case where the activity measuring apparatus 92 has detected an activity value in the dust of the small collecting container 89 that exceeds the predetermined limit value. The powder is directed with the aid of a two-way valve 98 towards the waste container 97. The valve in the two-way valve 98 directs the powder out of the small collection container 89 towards the waste container 97. In one embodiment, the 98 two way valve is manufactured by EMDE company, Zierenberg, Germany. Since only the least amount of dust that is most active is directed to waste container 97, only a smaller total amount of more activated dust is also generated, which can be stored more safely and economically. This more activated powder is subsequently packaged in Konrad containers and transported to final storage.

In step 70 of the flowchart (Figure 8), the powder 87 and 91 is loaded from the small collection container 89 into a waste container 99 in case the measured activity value does not exceed the predetermined limit value within time. predetermined. This waste container 99 is intended for released material that can be disposed of together with general debris. In this embodiment, the two-way valve 98 transports the classified powder from the small collection container 89 to the waste container 99, provided that the activity measuring apparatus 92 detects a probability of response of less than 7000 actual hits in the 7 kg of powder of the small collecting container 89 during the measurement interval lasting 10 seconds.

In step 71 of the flow chart (FIG. 8), the small collecting container 89 is again filled with powder 87 and 91 which falls from the dust separator 81 after the powder has moved either to the waste container 97 or either to waste container 99 in steps 69-70 of the flow chart. Each of the small collecting containers 88 and 89 is introduced into the activity measuring apparatus 92 after it has been filled with powder. After the expiration of the predetermined time interval, the powder may fall through the bottom of the small collection container into the two-way valve 98, through which the powder falls into either the waste container 97 or the container. debris 99. In one embodiment, the bottom of the small collection container can be opened to drop dust through the opening in the bottom of the activity measuring apparatus 92 and into 98. As soon as the small collection container 88 or 89 is Vacuum, it is removed from the activity measuring apparatus 92 and positioned again under the dust separator 81, where the small collecting container is again filled with powder. In another embodiment, the system for separating radioactively contaminated material 73 contains a second detector 100, which surrounds the ceramic recoil spark plug filter 90.

The smallest dust particles, which accumulate on the inner wall of the ceramic recoil spark plug filter 90, arrive from the suction opening 79 in the ceramic recoil spark plug filter 90 before the larger dust particles fall from the the bottom opening of the dust separator 81 and long before the small collection container 88-89 fills up and the activity of the small collection container is measured. Therefore, measurement of powder activity 91 in ceramic recoil spark plug filter 90 provides an early indication that radioactively contaminated powder reaches dust separator 81 and falls into small collection container 88-89. At least 100 grams of dust is assumed to accumulate in the ceramic recoil spark plug filter 90, i.e. the default limit value of 0.1 Bq / g for detection of radioactively contaminated dust corresponds to an actual hit rate of at least 600 within the predetermined time period of 60 seconds. In one embodiment, all cobalt-60 fragmentations are counted within a continuous time period of 60 seconds. In another embodiment, the predetermined time period is a 60 second time frame. In each second, the second detector 100 determines if at least 600 actual hits have occurred in the last 60 seconds. If the activity level, which has been measured by the detector 100 within the predetermined time period of 60 seconds, exceeds the limit value of 600 actual strokes, then the small dust-filled collecting container 88 or 89 is immediately emptied into the container debris 97 for radioactive waste without waiting for the completion of the filling of the small collection container. Then, the filling of the small collecting container begins again and continues as long as the predetermined limit value is not exceeded within a correspondingly predetermined period of time. The activity measurement of detector 100 is periodically interrupted during the phase recoil while cleaning ceramic recoil spark plug filter 90.

The system for separating radioactively contaminated material 73 also includes a safety filter 101, which is arranged after the ceramic recoil spark plug filter 90 to trap particles that may have passed through the ceramic recoil spark plug filter 90. The safety or emergency filter 101 rarely needs to be changed and can be disposed of in the waste container 97 without decisively increasing the amount of radioactive waste.

In turn another embodiment, the fixed period of time during which the activity measuring apparatus 92 measures the activity of the powder 87 and 91 is reduced to 1 second to enable a more constant flow and a continuous separation of the contaminated powder from the released powder. . The release limit value is maintained at 0.1 Bq / g. Instead of counting an equivalent adjusted by the response probability of at least 7000 hits within 10 seconds, the release limit value increases when the activity meter 92 detects more than 700 actual hits of the 7 kg of powder in the small collection container 88-89 within the predetermined time period of 1 second.

List of references used

1 Lock area

1.1 Exterior lock

1.2 Inner lock

2 High pressure press

3 Barrel drying / measuring installation

4 Konrad Container Loading

5 Exclusion zone warehouse

6 Installation of centrifuge wheel

7 Manual shot blasting box

8 Wet decontamination with high pressure water jets

9 Ultrasonic bath

10 Wet decontamination with PHADEC installation

11 Box washing facility

12 Thermal Fragmentation (28)

13 Manual Fragmentation (28)

14 Sheet Metal Processing (28)

15 Large component fragmentation (such as large component tent) (28)

16 Bandsaw (large)

17 Cable treatment (28)

18 Conditioning of asbestos (28)

19 Classification tent for radioactive waste (28)

20 Classification tent for non-radioactive waste (28)

21 Concrete crusher (28)

22 Concrete surface treatment (28)

23 storage and construction technology

23.1 Cleaning and service room

23.2 Oil and liquid warehouse

23.3 Stacker truck loading surface

23.4 Locker for tool distribution

23.5 Showcase for radiation protection

23.6 Office space

24 Buffer Areas

25 Solid walls in the building

25.1 Loading molten material

25.2 Cobalt Coincidence (CC) Measurement

26 Pre-measurement area

27 Building

27.1 Decontaminated Building (Interior) Stratification

27.2 Entry and exit for people

27.3 Emergency exit

27.4 Output for material that gives liberatory measurement (material that is below the radiological limit values according to the Radiation Protection Regulation)

28 Flexible wall

29 Transport routes

30 pieces of steel, contaminated

31 pieces of steel, giving liberatory measurement

32 Large components (steam generators)

33 Various wastes, contaminated

34 Radioactive waste, non-combustible

35 Radioactive waste, fuels

36 Non-radioactive waste, fuels

Mineral waste

Concrete slab, which gives liberatory measurement

Concrete rubble, giving liberatory measurement

Concrete rubble, contaminated

20 "container

Waste water treatment with evaporation system, concentrate collection container, reserve container, distillation container

Preparation of shot for blast cleaning

Hydraulic unit for high pressure press (2)

Pre-measurement area

Nuclear power station

Security container

Z-shaped concrete deck slabs

Pump Chambers

Pumps

I usually

Concrete floor

Crack

Decontamination varnish

Pollution

Covering

Cover layer, upper layer of the coating

Area from which uncontaminated material has been milled

Area from which contaminated material has been milled

Polishing concrete contaminated with radioactive particles to form dust containing the radioactive particles

Creation of a negative pressure that sucks the dust into a tube and then into a dust separator Filling the small collecting container with the dust falling from the dust separator

Detection of radioactivity in dust falling into the small collection container Transfer of dust from the small collection container to a waste container in case the measured radioactivity exceeds the predetermined limit value within a predetermined period of time

Transfer of dust from the small collection container to another waste container in case the measured radioactivity does not exceed the predetermined limit value within a predetermined period of time

Refilling the small collection container with powder falling from the dust separator after the powder has been transferred to a waste container

Vacuum pump

System for the separation of radioactively contaminated material

Manual concrete milling machine

Conveyor equipment

Chassis concrete milling machine

Oscillating chisel with hard metal (tungsten carbide)

Vacuum hose

Suction opening

Dust separator inlet

Dust separator

Cylindrical overlay

Air stream

Cylindrical lining

Tapered bottom

Bottom conical collection container

Powder

Small collection container

Small collection container

Ceramic recoil spark plug filter

Lumps of dust

Active measuring device, detector

Plastic Scintillation Detector

Fast photomultiplier

Plastic Scintillation Detector

Fast photomultiplier

Waste container

Two-way valve

Waste container

Detector

101 Safety filter (emergency filter)

Ċ

Claims (15)

1. Facility for the treatment of activated, contaminated and / or uncontaminated waste, in which, in a type of modular construction with low radiation exposure and low transport costs, activated, contaminated and / or uncontaminated waste or waste , decoupled from the decommissioning and regardless of a particular nuclear facility, are supplied to a recycling station, a landfill or a waste deposit after separation and / or decontamination, consisting of - a contact surface in a building (27) with an entrance and an exit (27.2, 27.3) for personnel, - at least one access for material, which is made as a lock (1, 1.1, 1.2) for contaminated parts, - at least one outlet (27.4) for material (31) to be delivered, which is decontaminated, - at least one outlet (1, 1.1, 1.2) for radioactive material or waste that can be released in a restricted way, - treatment stations (12, 13, 14, 15, 16, 17, 18, 19, 20, 21, 22 ), which are separated from each other by flexible and adjustable walls (28), so that their surface can be increased, reduced or modified with little effort according to the required performance, - treatment stations (2, 3, 4, 5), which are separated from each other by solid walls (25), as well as - transport containers, means of transport and storage spaces and rooms, the treatment stations being arranged , in which a high dose rate occurs, in an area with a high dose rate far away from the entry and exit of personnel (27.2, 27.3) and the exit (27.4) for decontaminated material (31) to be delivered. 2. Installation according to claim 1, characterized in that the building (27) is provided with an easily decontaminable covering (27.1). Installation according to claims 1 or 2, characterized in that the treatment stations comprise a large component fragmentation station (15), a thermal fragmentation station (12), a manual fragmentation station (13), a processing station sheet metal processing (14), a cable treatment station (17), an asbestos conditioning station (18), a sorting tent for radioactive waste (19), a sorting tent for non-radioactive waste (20) , a concrete crusher (21), a concrete surface treatment station (22), a box washing installation (11), a wet decontamination (8, 10), an ultrasonic bath (9), a dry blasting, a spinning wheel facility (6), a manual blasting box (7), a high pressure press (2), a Konrad container loading station (4) and a keg drying / measuring facility (3). Installation according to claim 3, characterized in that the concrete surface treatment station (22) has a concrete milling machine (74, 76), which is suitable for removing contaminated surfaces (58, 60, 61) from parts of concrete and to suck them by means of a vacuum pump (72) towards a dust separator (81) connected downstream of the concrete milling machine (74, 76), the dust separator being configured so that the predominant part is separated From a powder (87) down and the predominant part of an air stream (83) up and downstream of the dust separator (81) an activity measuring apparatus (92) is connected, which is configured so that it measures exposure to dust radiation (87). 5. Installation according to claim 4, characterized in that a filter (90, 101) is arranged between the dust separator (87) and the vacuum pump (72) to separate fine dust particles. Installation according to claims 4 and 5, characterized in that the activity measuring apparatus (92) represents a cobalt coincidence measuring apparatus, to which a two-way valve (98) is connected downstream, directing the powder ( 87) to a waste collection container (97) for contaminated material or a waste container (99) for released material. Installation according to one of Claims 4 to 6, characterized in that a cobalt coincidence measuring device is arranged as a detector (100) in the filter (90). 8. Procedure for the treatment of activated, contaminated and uncontaminated waste from the dismantling of nuclear facilities, treatment being carried out in a decoupled manner from the dismantling and independently of a specific nuclear facility with the aim of supplying the majority of the waste to a recycling station, the waste being supplied to an installation according to the characteristics of claim 1 and being treated in treatment stations, after which a decisive measurement takes place, and the decontaminated waste being supplied to a recycling station and the contaminated waste and radioactive waste to a nuclear cemetery. 9. Procedure according to claim 8 for the treatment of contaminated metal waste such as steel parts (30), characterized in that - in a first stage, the contaminated steel pieces (30) pass through a lock (1, 1.1, 1.2), in which they are registered and, if necessary, removed with suitable devices from the transport containers, - the contaminated steel parts (30) are supplied to a manual fragmentation station (12) and / or to a sheet metal processing station (14), where they are mechanically and thermally fragmented, - the contaminated (30) and fragmented steel parts are supplied to a wet decontamination station (10) or to a high pressure cleaning decontamination station (8), to a dry blasting procedure station, a procedure station cavity blasting machine, a spinning wheel process station (6) or a shot blasting box process station (7), - the treated parts are supplied to a decision measuring station, after which the contaminated steel parts (30) leave the facility through the sluice (1, 1.2, 1.1) and the releasing measurement material leaves the facility through the outlet for material that gives liberatory measurement (27.4). 10. Method according to claim 8, characterized by a treatment of mineral residues, such as rubble, by - a previous measurement of the rubble - a division as conventional or radioactive waste - screening and coarse shredding of conventional waste with subsequent pre-measurement and decisive measurement and packaging of conventional waste for a landfill or recycling station, and - screening and coarse grinding of radioactive waste, and - a filling in a container for a warehouse, and - an evacuation of waste. 11. Method according to claim 8, characterized by a treatment of waste, such as cables, by - decontamination of the cables by rubbing or other suitable procedure - a stripping of the cables - a crumbling of the cables and the stripped parts - a decisive measurement of the peeled and crumbled parts of the cable, after which - packaging and recycling of reusable waste, and - a filling of contaminated waste in a waste container with subsequent transport and storage. 12. Procedure according to claim 8, characterized by a treatment of waste, such as waste, by - a classification of the waste into conventional and radioactive waste, the conventional fuel waste being pressed and supplied to a pre-measuring and decision-making station, after which the uncharged waste is supplied to an external conventional combustion plant, and - radioactive waste is separated into combustible and non-combustible waste, the combustible radioactive waste being pressed and supplied to a dose rate and contamination rate station and an external combustion plant, and the radioactive non-combustible waste being supplied to a press high pressure and to a drying station, and - radioactive waste, burned and pressed and dried, is filled into a container or other suitable container and is evacuated and stored. 13. Procedure according to claim 8 for the treatment of mineral residues, such as concrete pieces, characterized by - transportation of mineral waste (37) through the lock area (1, 1.1, 1.2) to the concrete treatment facility (22), - where appropriate, dismantling of large concrete pieces and removal of the pieces that hinder the measurement of contamination, - a surface treatment as well as a detachment of the surface for decontamination, - a preliminary measurement and a decisive measurement, - an allocation to conventional waste and residues that can give a release measurement or radioactive waste, and - transport of contaminated concrete rubble (40) to the Konrad container loading station (4) and further on to a nuclear cemetery, - transport of a concrete slab (38), which can give a liberatory measurement, through the lock area (1, 1.1., 1.2) outside the installation, and -Transportation of concrete rubble (39) that gives liberatory measurement, through the outlet (27.4) for the material that gives liberatory measurement, outside the installation. 14. Method according to claim 13, characterized by - detachment of a contaminated surface (58, 60, 61) from a concrete piece and dust extraction (87) resulting by an air stream (83) towards a separator (81), - separation of the powder (87) and the air stream (83) in the separator (81), - separation of fine particles of residual dust from the air stream (83) by means of a filter (90, 101), - determination of the radioactivity of the dust particles (87) after the separation process with an activity measuring apparatus (92), - transfer of the powder particles (87) to a waste container (97) for contaminated material when the measured radioactivity exceeds a predetermined limit value within a predetermined period of time, - transfer of the powder particles (89) to a waste container (99) for material giving liberatory measurement when the measured radioactivity does not exceed a predetermined limit value within a predetermined period of time. 15. Method according to claims 13 and 14, characterized in that the radioactivity of the powder in the filter (90) is measured according to the cobalt matching procedure.